ML20154S477

Safety Evaluation Supporting Amend 47 to License DPR-45

ML20154S477
Person / Time
Site:	La Crosse File:Dairyland Power Cooperative icon.png
Issue date:	03/27/1986
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20154S469	List: ML20154S466 ML20154S470 ML20154S477
References
NUDOCS 8604010256
Download: ML20154S477 (8)

Text

Q TEC UNITED STATES 8' I e NUCLEAR REGULATORY COMMISSION r

g p WASHINGTON, D. C. 20555

\...../

SAFETY EVALUATION BY TbE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO.47 TO PROVISIONAL OPERATING LICENSE NO. OPR 45 DAIRYLAND POWER COOPERATIVE LA CROSSE BOILING WATER REACTOR (LACBWR)

DOCKET NO. 50-409

1.0 INTRODUCTION

By apolication dated December 12, 1985, Dairylar:d Power Cooperative (DPC)

(the licensee) requested a change to the technical specifications of the la Crosse Boiling Water Reactor (LACBWR). The requested amendment furnished information to support authorization for LACRWR to install replacement control rods of the ASEA-ATOM (AA) plate design with horizontally drilled absorber holes in place of the currently used Allis-Chalmers (A-C) tube-sheath desion control rods. The change would allow use of control. rods of both designs in the reactor. The requested change would also delete the requirement to go-gage control rods, as contained in Technical Specification 4.2.4.10, based on LACBWR and industry operating experience and because this surveillance is not contained in Standard Technical Specifications for Boiling Water Reactors (BWR).

2.0 EVALUATION 2.1 Evaluation of the Use of the AA Control Blades 2.1.1 General Description of the Blade The AA control blade for U.S. BWR is described in detail in the Topical Report TR-UR-85-225, "BWR Control Blade for US RWR's" submitted to the NRC in October 1985 which has been approved by the NRC staff. The AA control blade (rod) described in the topical report has been slightly modified in order to fit the LACBWR core lattice geometry, to closely match the reactivity worth of the original (A-C) LACRWR control rods, and to be compatible with the LACRWR control rod drives and control rod handlino equipment. The total weight in air of the control rod (includina extension) is calculated to_be 74 kg (163.16 lbs). Thus, it is slightly lighter than the A-C control rod which weighs approximately 77 kg (170 lbs.).

2.1.2 Materials Comoatability The cruciform absorbers section of the AA control blade is formed by welding four solid stainless steel plates together at the center. High purity (reduced Si, P, N and Co impurities) Type SS 2352-28 (AISI 304L) stainless steel is used for these plates. The intermittent center weld- ,

i 8604010256 860327 ADOCK 0 409 gDR ,

foint ensures straightness and required stiffness while permittina 17 cut-out sections which results in significant weicht savings. Thus, the total weight of the control rod is slightly less than that of the original A-C rod, in spite of the slightly heavier absorber-containing part of the AA rod.

The rod wings are 8.05 mm (0.317 in.) thick. The absorber section of each wina has 245 horizontally drilled 6 mm (0.236 in.) diameter holes filled with B C power and 19 similar holes containing Hf-pins. The holes 4

are spaced at a pitch of 8 mm (0.315 in.) resultino in a total absorber section length of 7110 mm (83.071 in.). The holes are 104 nun (4.09 in.)

deep. The absorber part of the control rod coincides with the active core height when the control rod is full inserted. The top 101 mm (3.976 in.)

of the rod is free of holes. This section constitutes a " grey" nose which reduces local power chances, thus mitigating fuel duty.

The horizontal holes are filled with natural B 3C by vibratory compaction to a packing density of 70 3% of the theoretTcal density. Each control rod contains about 5.2 kg (11.5 lbs.) of B 4C plus 2.9 kg (6.4 lbs.) of hafnium.

The holes are closed at the outer blade edge but are connected throuch a narrow slit. This design allows gas pressure equalization between holes but prevents any significant displacement of the B C powder. The horizontal holes also render any further 3B C densification after initial filling quite insignificant. Each wing forms a seoarate pressure enclosure which is pressure and leak tested after welding.

A lifting handle, cut frnn the same materf al as the control rod wings, is welded to the top end of the absorber section. This handle is designed to fit the grapple used for installing the control rods, and corresponds to the A-C handle desian. At the too end of the rod are cuide; pads, one on each wina, which prevent direct contact between the control rod and the adjacent fuel shrouds. The guide pads are made of Inconel X-750 which contains less than 0.1 weight-% cobalt. The control rod absorber section and the extension at the bottom are welded together. The extension corresponds to the previous desian for LACBWR control rods.

The AA control rod is desianed for an internal overpressure of 15 MPa (2175 psi). The internal gas pressure is a function of the average boron depletion in a rod wing. Due to pressure equalization, all absorber holes in a wing are under tha same pressure. The hafnium tip control rod will have a slightly lower gas pressure buildup than an all B C g rod due to the reduced axial burnup variation in the B C 4 and no gas release from hafnium. During neutron exposure, gas pressure will build up mainly due to helium production .

from neutron capture in 8-10, and as a result of radiolysis of any water l traces present. The maximum moisture content of the B 4C is specified as 350 ,

ppm at filling in order to limit the water pressure contribution. Other reaction pro 6ucts in the gaseous form may be neglected. Gas pressure buildup is not expected to be life-limiting for this control rod design.

I

The control rod is provided with a 150 mm (5.9 in.) hafnium tin. The uopermost 19 holes in each wing are filled with hafnium metal rods instead of B,C powder. The hafnium is of reactor grade with 4.5 w/o zirconium.

Hafnium exhibits no irradiation induced swelling and proper tolerances ensure that the hafnium will not induce any stresses in the stainless steel plate during irradiation. The use of hafnium in control blades has oreviously been approved for test blades in Peach Bottom and Dresden Unit 3, and is an alternate for the silver-indium-cadmium (Au-In-Cd) used in Westinghouse reactors. The staff is unaware of any materials problems associated with the use of hafnium and finds this aspect of the design acceptable.

2.1.3 Nuclear Desian Characteristics i

The nuclear design characteristics of the improved AA control blades have been performed by AA with the PHOENIX lattice and depletion code.

The code has been used to compare reactivity worths at cold xenon-free conditions and hot voided and unvoided conditions as a function of fuel burnup. In addition power distribution effects and absorber depletion effects have been studied. The conclusions of the analyses are discussed below.

The reactivity worth of a control rod is directly dependent on the rate of neutron absorption in the rod materials-9 C/Hf and stainless steel. The 3

total absorotion rate depends on several factors: 6 mount and spatial distribution of the materials in the rod, and the neutron energy spectrum which in turn depends on the core conditions (temperature, void fraction, fuel assembly type and its exposure history). Eventually, the decletion-of the highly neutron absorbing material will reduce the absorption rate, and hence, the control rod reactivity worth.

The reactivity worth of the AA control rod for the LACBWR relative to the original A-C control rod was calculated with the two-dimensional lattice depletion code PH0ENIX for different core conditions. From a neutronics point of view, the design of the AA control rod differs from that of the original LACBWR rod mainly by its 35% larger volumetric inventory of neutron absorber (B C/Hf). As a result of this difference, the B C 4 region of the AA control rod 4has a consistently higher reactivity worth, for the all rods inserted condition, of about 2.6-3.0% (relative) for the various reactor conditions. Cctresponding values for the AA control rod tip (5.9 in.)

with Hf are 5.0-5.9% (relative) less worth than that of the tip of the A-C rod. The overall reactivity worth of the AA control rod, in the all rods inserted condition, is about 1.3-1.8% (relative) preater than that of the

existing A-C control rod for the various reactor conditions. In the hot, fulloower, 20% moderator voio ?se with 75% or more of the control rods withdrawn from the core (aver, i aperating condition) the relative reactivity worth of the AA control rod is ).0-0.3% less than the A-C rod.

The increased neutron absorption rate in the AA control rod is essentially entirely due to the volume absorption of epithermal neutrons in B 4C. Other differences between the two control rod designs, such as the amount and distribution of stainless steel, have minor influence on the reactivity worth differences.

The models used in the PHOENIX two-dimensional calculations represented either the cross section of one fuel assembly or a control cell with or without aporopriate control rods inserted. The physical and neutronic properties of the LACBWR Type III fuel and other core components were accurately represented. Reflective boundary conditions rendered the calculated reactivity cuantities valid for an irfinite core. For the purpose of assessing the reactivity worth of control rods, especially with resoect to relative differences, calculations with PH0ENIX will oive reliable results. Calculations were performed for the followina reactor conditions using fresh xenon-free fuel:

a) Hot Full Power (HFP), zero void, (i.e., near the core inlet) b) HFP, 20% and 40% void c) Cold, clean, zero power (shutdown condition).

The dependence of the relative worths of the AA and A-C rods on xenon and fuel burnup was also studied and the effect of these parameters was found to be ouite negligible.

The difference in neutron absorption rates in the various control rod designs will have a small effect on the neutron flux density distribution in fuel assemblies near the control rods. The greater neutron absorption rate in the B gC region of the AA control rod, compared to the A-C control rod, causes a slightly areater suppression of the power density in fuel rods near the AA control rod and a slightly areater power peaking in fuel rods away from the AA control rod. In the two dimensional infinite lattice model, the maxium increase in fuel rod power density (relative to that with A-C control rods) is less than 0.05 kW/ft for the hot, full power, 20% void, no xenon Case.

For the AA control rod Hf-tip section, the suppression of the power in the fuel near the control rod is less than that for an A-C control rod. The power density distribution in the fuel in the vicinity of the Hf region of AA control rods is actually flatter than with A-C rods and peak fuel power densities are lower.

I

- - - , , .- - - - ,%. v ., . - . , , ,

i l

The burnup of the neutron absorber in a control rod will lead to decreasing neutron absorption and hence, reduced control strength. The rate of reactivity decrease as a function of neutron exposure depends on the initial absorber inventory (i.e., on relative effective consumption of absorber nuclides). With about 50% areater B4 C content per unit length than the A-C control rod, the AA control rod will have a corresponding exposure time advantage for any given relative reactivity decrease or relative depletion of B-10. The reactivity worth of the hafnium recion of the AA control rod will decrease even slower than the BgC region. This is due to the fact that natural Hf consists of several isotopes, where the most important are Hf-177, Hf-178, Hf-179 and Hf-180. All these isotopes have significant absorption cross-sections An absorbed neutron will give rise to a new Hf isotope which in turn may absorb a neutron.

For negative reactivity insertion events (scram events) the slightly greater reactivity worth of the AA control rods is expected to be beneficial. Slightly more negntive reactivity will be inserted faster than with the A-C control rods. The shutd iwn margin will also be slightly greater with AA control rods than with the u.crent A-C control rods.

For positive reactivity insertion events, i.e., inadvertant control rod withdrawal and control rod drop accident, the slightly greater worth of the AA control rod is also expected to have a minimal effect. The limiting anticipated transient for the LACBWR is the inadvertent control rod withdrawal at operating power. A licensee recalculation of the limiting control rod withdrawal transient for the beoianing of Fuel Cycle 9 using a conservative 3% frelative) creater control rod worth resulted in a Minimum Critical Power Ratio (MCPR) of 1.529 compared to a MCPR of 1.539 calculated for the A-C rods. The effect produced by the actual LACBWR AA control rod with a relative worth equal to or slightly less than the A-C rods in the operating reactor would be completely negligible. In the LACBWR, the consequences of control rod drop accident are greatest when the reactor is at power. The results of the probability study of the LACBKR control rod drop accident are not very sensitive to the specific worth of the control rods and the small differences between the AA and the A-C rods would have an insignificant effect on the results.

2.1.4 Control Rod Maneuverino The AA control rods are essentially identical in exterior envelope to the A-C rods. The AA control rod is slightly lighter (approximately 7 lbs.) than the A-C rod and therefore, scram times for the AA rod should be approximately the same as or slightly faster than for the A-C rod. Scram times for the AA control rods will be measured after installation in the reactor as required by current procedures.

2.1.5 Blade Experience and Surveillance Procram AA rodshas accumulated of similar designmany(years of operating except for the Hf tip) to experience the LACRWR withAA some 1200rods.

control control These control rods have performed very well in the AA BWR. Only after 8 to

12 fuel cycles were a few instances of interarar:ular stress corrosion cracking (IGSCC) observed along 4R C absorber holes near the top of control rods where the B-10 depletion was greater than 60%. Hot lab destructive examination and neutron radiography of several control rod segments with IGSCC showed that B 4C washout necurs in the hole with a crack but only to a limited extent in neighboring holes due to the gas pressure eoualization passages. These examinations also showed that a simple visual in-pool inspection can detect almost all cracks, thus a visual inspection program can be used to check the integrity of AA control rods. The improved AA control rods with Hf tips and high purity 304L SS should have a considerably longer failure free life than the rods in the above exanination program. Based on this background, AA currently recommends the following surveillance guidelines for AA Hf tio control rods:

Control rods which show no visible cracks can be used without any restrictions for another fuel cycle.

Control rods which show visible cracks should be replaced.

The licensee has proposed to crudently base control rod lifetime on exposure histories, R-10 depletion, and visual examinations as is currently the case for the rest of the U.S. RWR.

Based on the above and the use of hafnium tips and high purity 304L SS, the staff has concluded that there is not a cracking-related safety concern from use of AA rods in the LACRWR.

2.1.6 Sumary On the basis of its review the staff has concluded that the use of the AA improved control blades in LACBWR is acceptable. This conclusion is based on the following considerations:

1. The improved blades are mechanically and hydraulically compatible with the present control blades.

2. The nuclear characteristics of the blades have been determined by acceptable methods.

3. The effect of the AA rods will be insignificant or slightly beneficial during transiet events.

4. Sufficient experience exists with the rod design in other (Swedish and.

Finnish) BWRs to permit the conclusion that they will- operate without significant deterioration.

5. A satisfactory control rod lifetime program has been established to monitor the blade performance.

6. Hafnium metal is used for neutron absorption in the top 5.9 inches of the absorber region. Hi does not deplete as fast as B-10 in B C and does nct swell with irrediation and, therefore, effectively eliminakes the very significant problem of high local B-10 depletion in the tip of an all B 4C rod like the A-C rod.

7. There is approximately 50% greater B C content per unit length in the 8 4C region than the A-C rod. Therefore,4a correspondingly longer exposure tine is required for a given decrease in reactivity worth or relative B-10 depletion.

8. The blades are made of high purity (refcced levels of Si, P, and S impurities) Type 304L stainless steel which has a lower susceptibility to IGSCC than ordinary 304L stainless steel.

2.2 Evaluation of the Deletion of the Surveillance Requirement to Go-Gaae

' Control Pods (TS 4.2.2.10)

The requirement to geoe the control rods was originally based on the belief that the life of the rod would be limited by the pressure buildup in the R4 C tubes due to helium release from the irradiated BgC and that the gaae would detect tube swelling before tube failure. Experience has shown that absorber tubes can fail by IGSCC and B 3C be lost be# ore any swellino is detected by gaging. The licensee has never detected swelling of a control rod blade even with 0.470 inch go gages. The industry has concluded, after extensive research, that absorber tube failure by IGSCC and R4 C swelling is more a function of R-10 depletion, and resultant B C 3 swelling and change in physical characteristics, than a result of pressure buildup from helium release. The requirement for more frequent inspections after reaching an exposure of "12,800% delta k/k MWD /MT" was also based on the theory of pressure buildup due to helium release and not on any limitation due to reduction of reactivity worth or B-10 depletion.

In lieu of the requirements in the current Technical Specification 4.7.4.10, the licensee proposes to continue to follow the exposure history and B-10 depletion of each control rod in the LACBWR and base a prudent control rod examination and control rod shuffle and/or discharge program on this data.

Based on the above, and since this surveillance requirement is not contained in Standard BWR Technical Specifications, the staff finds deletion of this specification acceptable.

3.0 ENVIRONMENTAL CONSIDERATION

This amendment involves a change to a requirement with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20 and changes to the surveillance requirements. The staff has determined that the amendment involves no

significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure.

The Commission has previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public comment on such finding. Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement nor environmental assess-ment need be prepared in connection with the issuance of this amendment.

4.0 CONCLUSION

The staff has concluded, based on the considerations discussed above, that:

(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security nor to the health and safety of the public.

Principal Contributor: J. Stang Dated: March 27,1986 i

l l

.CIM _